HBL-05-010, ATI Consulting Report Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant

From kanterella
Jump to navigation Jump to search
ATI Consulting Report Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant
ML051170442
Person / Time
Site: Humboldt Bay
Issue date: 03/31/2005
From: Montgomery R, Nickell R, Rashid Y, Server W, Sunderland D
ANATECH Corp, ATI Consulting
To:
NRC/FSME
References
PG&E Letter HBL-05-010
Download: ML051170442 (34)
v • d • e


Text

-

I Enclosure 1 PG&E Letter HBL-05-010 ATI CONSULTING REPORT EVALUATION OF NUCLEAR FUEL ROD FRAGMENTS AND INFERENCE TO FUEL ROD A-49 AT HUMBOLDT BAY POWER PLANT DATED MARCH 31, 2005

i EVALUATION OF NUCLEAR FUEL ROD FRAGMENTS AND INFERENCE TO FUEL ROD A-49 AT HUMBOLDT BAY POWER PLANT Revised version reflecting review comments and additional information from PG&E staff by ATI Consulting March 31, 2005 Expert Team Members:

William L. Server Dr. Robert Nickell Dr. Y. R. Rashid, Robert Montgomery, and Dion Sunderland, ANATECH Approved by: ________ Date: 31 March 2005 William L. Server, President, ATI Consulting

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 EVALUATION OF NUCLEAR FUEL ROD FRAGMENTS AND INFERENCE TO FUEL ROD A-49 AT HUMBOLDT BAY POWER PLANT Executive Summary Still photographs, drawings, and digital video disks (DVDs) have been extensively reviewed by an expert review team assembled by ATI Consulting. The judgment of this team, after reviewing all available information, has concluded that there is reasonable evidence consistent with the proposition that fragments from the three 18-inch-long segments along with remnants cut from the A-49 fuel rod may be amongst the fuel fragments in the Humboldt Bay Power Plant spent fuel pool.

1

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005

Background

The Humboldt Bay Power Plant (HBPP), Unit 3, near Eureka in northern California, was shut down in 1976 after some thirteen years of operation. The plant owner, Pacific Gas & Electric Company (PG&E), is preparing to move the fuel from the spent fuel storage pool into dry storage casks for long-term storage at an independent spent fuel storage facility (ISFSI). Since late 2003, plant personnel have been characterizing and documenting the contents of the spent fuel storage pool in preparation for the ISFSI transfer. On June 23, 2004, a discrepancy in the plant records was discovered relative to one fuel rod from Assembly A-49. The records showed that the entire assembly had been shipped to a fuel reprocessing facility in West Valley, New York, on August 6, 1969. However, other records showed that, prior to the West Valley shipment, one fuel rod had been cut into segments and removed from the assembly. Records indicate that three approximately 18-inch-long segments were cut from the fuel rod. The end pieces and other remnants of the fuel rod, in whole or in part, were either placed in a storage container in the spent fuel pool or were shipped with assembly A-49 to West Valley. The three 18-inch segments, while prepared for shipment to Battelle-Columbus Laboratories (BCL) for study, were not shipped and the last written record located to date indicates that a 20-inch long pipe containing the three 18-inch segments was hung from the side of the spent fuel pool (SFP) in late 1968.

BCL records confirm that the three 18-inch segments were never received. PG&E notified the Region IV office of the U. S. Nuclear Regulatory Commission (NRC) concerning the potential discrepancy on June 28,2004, and began the process of reconciliation of the records. On July 7, 2004, a physical search for the three 18-inch-long cut segments was initiated. Potential locations within the HBPP spent fuel pool, including the pool walls and storage containers were examined without discovery of the intact missing fuel segments or the steel pipe container (stainless or carbon). However, significant numbers of fuel fragments were located in storage containers in the pool and, in some cases, on the bottom ofthe spent fuel pool. More than enough fuel fragments relative to those needed to account for the missing inventory were located, but the sizes of the fragments and the initial examination for fragments with cut ends did not exhibit the expected characteristics for the fuel rod segments removed from Assembly A-49 (henceforth, often referred to as the A-49 rod). PG&E's search of records and plant personnel interviews have indicated that the 1968 segmentation of the A-49 rod represents the only time that a fuel rod had been mechanically cut at HBPP, thus making the location of a mechanically cut end the prime identifier (other than stainless steel cladding) of a fragment that originated from the A-49 rod.

In an effort to resolve the issues surrounding the fuel fragments, PG&E contracted with ATI Consulting (William Server) to review digital video and still photographs of the inventoried fuel rod fragments to determine their potential to be part of the remnants and cut segments from the fuel rod from Assembly A-49. ATI Consulting has enlisted the help of Dr. Robert E. Nickell as a technical consultant and the professional services of ANATECH Corporation (Dr. Y. R. Rashid, Robert Montgomery, and Dion Sunderland). These individuals have extensive experience with nuclear fuel issues and have some knowledge of the type of fuel/cladding used in the early days at HBPP. Resumes for all of these individuals have been provided to PG&E.

2

l ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Sources of Data Evaluated The initial sources of information and data reviewed and evaluated by the examiners were:

  • A 21 minute, 15 second video footage on DVD entitled "Fuel Fragments with possible cut end/ends"
  • An MS Word document containing digital pictures entitled "Fragment Picture Contact Sheet"
  • Three pictures of FF013 (fuel fragment # 13) that were attached to the above MS Word document
  • Two e-mail reports by Will Barkhuff (PG & E metallurgist) summarizing his assessment of fuel fragments in the HBPP spent fuel pool and other data
  • Telephone conversations involving Bruce Norton (HBPP Special Nuclear Material Inventory Project Manager), Will Barkhuff, Robert Nickell, and William Server (ATI Consulting)
  • Background information from a presentation made to the NRC concerning key aspects of this issue was provided by PG&E in MS PowerPoint file "Sep29'04NRCMtg9-27draft3".

A preliminary review was conducted in mid-December 2004 based upon this above information.

It was obvious that additional information was needed in order to finalize a position relative to the fuel fragments. After a site visit to HBPP by Dr. Nickell on December 22, 2004, additional digital videos and related information were obtained and reviewed that allowed a more complete evaluation to be performed. Some of this additional information was received as late as March 2005 after a thorough technical review of a draft version of this report by PG&E staff. Key items that were included in the evaluation are listed below:

  • HBPP Onsite Review Committee - Minutes of Special Meeting, September 17, 1968.;

this document describes plans for cutting the A-49 fuel rod to produce three 18-inch segments for shipment to BCL (four cuts are identified); this is termed the "original cutting plan" for this report.

  • Drawing showing exposure of A-49 corner rod as a function of distance from the top of the rod; this drawing includes a proposed cutting plan that would require five cuts to produce three 18-inch segments and is termed the "alternate cutting plan" for this report.
  • Fuel assembly A-49 drawings and notes showing and describing the operational damage to fuel rods in A-49, dated September 7, 1966.
  • Design sketch for fabrication of the pipe container for the three 18-inch fuel rod segments.
  • HBPP Onsite Review Committee - Minutes of Regular Meeting, October 2, 1968; this document describes the cutting of the fuel segments, placing of the unused portion of the rod in the central storage container in the SFP, cancellation of the shipment to BCL, and storage of the container of fuel segments in the SFP.
  • HBPP Unit 3, Quarterly Operations Report No. 68-3, July, August, and September 1968.
  • HBPP Unit 3, Technical Memorandum No. 66-1, Detection of Fuel Cladding Defects by "Sipping" Technique During September 1965.

3

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 R. N. Duncan, et al, "Stainless-Steel-Clad Fuel Rod Failures," General Electric Company, Nuclear Applications Vol.1 (pp 413-418), October 1965.

The following discussion presents the general assessment of the expected condition of the fuel rod segments removed from the A-49 rod, the evaluation criteria used to assess the visual appearance of the ends of the fuel rod fragments recently examined in the HBPP spent fuel pool, and the details leading to a final assessment of the fuel rod fragments with respect to the cut 18-inch long segments from rod A-49.

4

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 General Assessment During HBPP Cycle 1-B, the GE Type 1 fuel assemblies designed for use in HBPP began to experience fuel rod failures due to intergranular stress corrosion cracking (IGSCC) in the stainless steel cladding. IGSCC-induced cladding failure in Type 1 fuel rods was also observed at that time in the Dresden and Vallecitos Boiling Water Reactors. An inspection campaign performed following the completion of HBPP Cycle 1-B found a total of 32 leaking fuel assemblies and 16 suspect leaking fuel assemblies.

Figure 1 shows the dimensions of a typical Type 1 fuel rod representative of the stainless steel clad fuel rods in Assembly A-49. The axial dimension, not including the endplug shanks at the top and bottom ends of the rod, is approximately 83 inches. The top endplug shank is slightly longer than one inch, while the bottom endplug shank is slightly longer than 0.5 inches. Based on the length measurements of the various fuel fragments, including any potential remnants from the cutting operation and the endplug shanks, this assessment assumed a total length of approximately 85 inches. Each fuel rod has an identification stamp at the bottom end of the rod above the endplug shank.

A description of the original cutting plan for Assembly A-49 (Figure 2) is contained within the minutes of the On-Site Review Committee (OSRC) meeting held on September 17, 1968. The original cutting plan called for inserting the assembly within a special channel containing notches (or cutting guides) and placing the assembly horizontal onto an inspection platform. It was noted in the plan that the comer rod was bowed and cracked in the area between the first and second grid spacers (see Figure 3) due to operation within the reactor and that care should be taken when placing the assembly within the special channel to ensure the comer rod fit inside the channel.

No information has been identified that describes the characteristics (or dimensions) of the special channel other than the references to its use in the original cutting plan. Once placed on the inspection platform, the original cutting plan called for the use of an underwater hacksaw to cut the rod. However, it should be noted that interviews with plant personnel have found that a radial grinder may have been used to perform the cutting operation. Following the cutting operation, each cut segment was to be placed into a steel container in preparation for shipment to BCL. Finally, the remnant pieces were to be removed from the assembly if they were loose and placed in the SFP central storage container.

A schematic of the original cutting plan for the fuel rod in Assembly A-49 is shown in Figure 2 and consisted of the removal of 54 inches of the fuel rod as three 18-inch long segments. Prior to cutting, the procedure called for identification marks to be placed at the top of each segment using a file. If the original cutting plan was followed, the removal of the three 18-inch long segments would have required a total of four cuts. The original cutting plan would produce two remnant pieces from the lower and upper sections of the rod: a short bottom piece approximately 1-inch long and a longer top piece approximately 30-inches long, including the endplug shanks.

Based on the original cutting plan, the fuel rod cut from Assembly A-49 should have been separated into five (5) pieces with a total of eight (8) cut ends.

During the review of the records related to Assembly A-49, a second (alternative) cutting plan was discovered that also appears to address the damaged portion of the rod between the first and second grid spacers and includes a fifth cut near the top of the rod. This alternative cutting plan is shown in Figure 3. Visual inspection of Assembly A-49 during the inspection campaign found 5

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 '

that the corner rod selected for cutting had sustained damage in the form of cracking and bowing in the region between the approximate 30-inch elevation and the approximate 40-inch elevation (see visual inspection notes, September 7, 1966). This damaged region corresponds to the second 18-inch long segment in the original cutting plan. The alternative cutting plan appears to be an attempt to by-pass the handling and shipment of the damaged region between the 30-inch and 40-inch elevations of the corner rod from Assembly A-49. In the alternative cutting plan, three remnant pieces would be produced: an 18-inch long central piece, an approximately 12-inch long top end piece, and an approximately 1-inch long bottom piece, including the endplug shanks. Based on the alternative cutting plan, the fuel rod cut from Assembly A-49 would have been separated into six (6) pieces with a total often (10) cut ends.

Comparing the original and alternative cutting diagrams with the Type 1 fuel assembly design identifies that the second cut from the bottom in both plans would be less than 0.5-inches away from the lower spacer grid. This comparison also is shown in Figure 3; an estimate of the spacer grid locations is based on the poor quality diagrams in the HBPP archive and other sources. The proximity of the second cut location to the spacer grid may have complicated the cutting process.

Furthermore, the third cut appears to be very close to the cracked and bowed region noted in visual examinations. Investigations of the IGSCC failures in the Dresden and Vallecitos BWRs noted that most cladding cracks occurred in the peak power area where the burnup was greatest and the cladding stresses would be the largest. The third cut at the approximate 38-inch elevation is at the peak burnup location and would have been within the part of the fuel rod with the most damage from IGSCC. The amount of localized embrittlement within this region would have made the cutting process difficult to perform without further damaging the fuel rod segments. The combination of these factors may have resulted in the selection of still other alternative locations for cutting the segments other than those shown in Figures 2 or 3.

Documentation of such further modifications to the cutting plan has not been identified.

However, evidence has been found that suggests the cutting of the A-49 rod did not progress as originally planned and the fuel rod may have broken at least once during the sectioning campaign. As a result, the lengths and number of remnant pieces may be different than planned, resulting in uncertainty in the number of cut-affected ends.

Based on this understanding of the pre-cutting fuel rod condition and the possible scenarios of the post-cutting condition, the assessment described below was performed to identify the presence of the remnants of the threel 8-inch long segments amongst the fuel rod fragments found within the SFP.

All elevations are relative to the bottom of the fuel rod.

6

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Evaluation Criteria In order to provide a rational basis for the assessment of the fuel rod segments, a set of complementary criteria was developed for distinguishing between ends of broken segments produced by cutting, as opposed to ends of broken segments produced by other failure mechanisms, such as IGSCC circumferential cracking or tearing. The set consisted of five criteria: (1) planarity/perpendicularity of the end surfaces of the individual fuel rod fragments, relative to their axial orientation; (2) appearance of the separated surfaces of the cladding, when viewed normal to the separated surface; (3) flushness of the separated surfaces of the fuel pellets with the separated surfaces of the cladding; (4) appearance of the separated surfaces of the fuel pellets, when viewed normal to the separated surface; and (5) degree of associated intergranular stress corrosion cracking adjacent to the separated surface.

For the first criterion, the planarity or perpendicularity of the end cross sections of the individual fuel fragments is presumed to be due to the cutting actions during the generation of the three 18-inch sections and their separation from the end remnants. However, an added complication to the interpretation of planarity or perpendicularity of the separated cross section is the possibility of a near-circumferential separation of the stainless steel fuel rod cladding caused by IGSCC, driven, in part, by a circumferentially uniform axial stress. In order to distinguish between planarity or perpendicularity caused by a cutting action or that caused by near-circumferential IGSCC, additional criteria were defined related to either separated surface reflectivity (see the second criterion, discussed below), surface irregularity indicative of cracking along the grain boundaries, or the proximity of the separated surface to other IGSCC, perhaps oriented in, or branching from cracks in, other directions (see the fifth criterion, discussed below).

It was not assumed that the end cross sections were perfectly perpendicular to the rod axis; however, based upon verbal information with respect to the cutting operations, it was assumed that the cutting actions caused essentially planar geometry over most of the cross section.

Allowance was made for the possibility that the end surfaces may have separated prior to complete cutting of the cross section. That is, mechanical separation of the remainder of a cross section was considered after the cutting action had separated a substantial portion of the cross section. That final mechanical separation was assumed to include the possibility of tearing of the cladding and separation of the fuel pellets along ceramic fracture surfaces.

The second criterion is related to the separated surface reflectivity. Any cut end surfaces of cladding for the fuel rod fragments were expected to be extremely reflective under good lighting and viewing conditions, typical of a mechanically-cut surface that involves some degree of transgranular separation. Any end surfaces viewed normal to the separated surface were examined with this criterion in mind. The reflectivity helps to distinguish between the potential for IGSCC, which is less reflective due to the more irregular grain boundary separation.

However, the potential for other types of mechanical damage to that cut surface, as the result of abnormal handling conditions or other trauma, was considered. Judgment was required in order to determine the minimum amount of reflection that was needed to make a cut end finding. For example, if the cut portion of the end surface was thought to be about 75 % (after which complete separation of the other 25% occurred by ductile tearing), and trauma to the end cross section eliminated approximately half of the reflectivity of the portion of the surface that was deemed to be cut, the residual reflectivity was determined to be sufficient to make a cut finding.

7

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 The third and fourth criteria concern the degree of consistency betWeen the separated surface of the cladding and the separated surface of the fuel pellet itself, and the reflectivity of that separated fuel pellet surface. The flushness of the separated end of the fuel pellet with the separated end of the cladding is an important consideration in this regard, especially when combined with the appearance of the separated end of the pellet. Cracked pellet surfaces are not reflective when viewed normal to the separated surface under good lighting and viewing conditions. However, the cut portion of a pellet at the separated surface is very reflective. These two criteria have to be used judiciously, since mechanical trauma to the separated ends can readily cause fuel pellet fragments to crack and spall off. In a number of cases, however, the flushness of the separated ends of the fuel pellets and cladding, when combined with the reflective appearance of the separated fuel pellet surface, were consider strong evidence of a cut surface.

Finally, the fifth criterion was applied in order to assist in distinguishing between nearly circumferential IGSCC and cutting. IGSCC often is characterized by a dominant cracking orientation and less dominant cracking, possibly in different crack orientations, especially when the stress field is not driven by a single stress component. In this case, the areas adjacent to the separated surfaces were examined to determine whether IGSCC was observed, thus implying that near circumferential IGSCC was a possibility. Further, since circumferential IGSCC occurs predominantly at the pellet-pellet interfaces, it would be easy to distinguish it from a cut by looking at the fuel reflectivity, (the pellet ends are generally far less reflective than a cut pellet),

and whether or not the fuel pellet is flush with the cladding.

8

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Data and Discussion An estimated mapping of some key fuel fragments: FF013, FF026 and FF021, and the possible location of the three 18-inch segments are also shown in Figure 3, all overlayed on a diagram of the "alternative" cutting plan. Two potential remnants from the cutting operation - designated FF021 and FF026 - are approximately 14 inches long and 11 inches long, respectively. Of these fuel fragment FF026, which clearly has an upper endplug at one end (based on the endplug shank length - Figure 4), meets the expected length of the upper most fuel section remnants based on the alternative cutting plan (see Figure 3). Furthermore, the fuel fragment FF026 satisfies all the criteria for a cut at the other end.

The end of FF021 appears to be a lower endplug based on the length of the shank (see Figure 5).

Unfortunately, the separated end of FF021 has been split and bent back, in a very ductile manner, possibly because of trauma. As a result, it is not possible to observe the remnant end for cut characteristics. One possible explanation for such trauma could be handling operations during placement of the remnants into the SFP central storage containers or in some other manner. It is not expected that the significant bending trauma to FF021 would have occurred during operation or the assembly cleaning/dechanneling processes performed during early 1960s. In any case, FF021 would be about one inch longer without the trauma (-15-inch total length). To our knowledge, one other fuel fragment with an upper endplug has been found. This fragment (FF009) is approximately 30-inches long, but it does not meet any of the criteria for having a cut end.

FF013 (Figure 6) represents a fuel fragment that appears to have an artificial feature consistent with a partial cut. The broken side of FF013 has the characteristics of a circumferential break with very rough features, rather than a clean, sharp cut. The other end of FF013 shows characteristics of ductile tearing. The shape of this end suggests the possibility that FF013 mated with the cladding of FF021 that was 'split and bent back'. Together, the combined length of FF013 and FF021 is approximately 18 inches and perhaps as long as 19 inches accounting for the folded and bent cladding.

It is reasonable to conclude, then, that this partial cut coincided with rod breakage precipitated by prior IGSCC damage. The presence of the groove mark in FF013, along with the two broken ends, suggests that the cutting campaign did not progress as planned and that adjustments may have been made in the remaining cutting operations (e.g. Cuts I and 2 may never have been made). For example, breakage of the fuel rod below Grid Spacer #1 would have allowed for removal of the lower piece (FF021) from the fuel assembly without the need for performing Cut#l in the cutting plan. Furthermore, it is possible that at least one of the 18-inch long segments may not have a cut end, but an end that exhibits fracture characteristics.

The combined length of FF013, FF021 and FF026 is about 29-30 inches, including the endplug shanks, leaving about 54 inches left for segments that were to be shipped to BCL. Three 18-inch-long segments would be about 54 inches. Based on the visual characteristics of the three fuel fragments (FF013, FF021, and FF026) and the expected difficulties in cutting a fuel rod from Assembly A-49, FF013, FF021, and FF026 may constitute remnants of the A-49 rod, and t Digital videos show examples of the removal of items from the spent fuel pool central storage container, and personal communications with HBPP personnel confirmed the potential for trauma of the ends of items during placement into the container.

9

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 the three 18-inch-long cut segments in combination could constitute essentially the entire fuel rod.

Considering the potential cut ends from FF026 and the break at the upper end of FF013, an additional five (5) cut ends are being sought from among the other fuel rod fragments that have been located in the HBPP spent fuel pool. With respect to Cuts #3, #4, and #5, the fuel rod may have been partially or completely broken at or near the location of Cut #3 due to the IGSCC damage during irradiation. This damage was noted in visual inspections of the corner rod in Assembly A49. As a result, Cut #3 may not have been necessary or only a partial cut may have been required to remove from the assembly the part of the fuel rod labeled Segment #1 in Figure

3. In the alternative cutting plan, Cuts #4 and #5 would have been performed on regions higher up on the rod that possibly had less IGSCC damage. Cut #5 would coincide with the end of FF026, and, based on the appearance of the separated end of FF026, it is possible this cut was made successfully. From these observations, the number of cut ends associated with the missing segments could be as many as five (5) or as few as three (3).

To date, other than FF021 no fuel rod fragments even remotely close to 18 inches in length have, been located in the search, nor are such fragments likely to be found at this juncture in the investigation. The fragments that have been examined so far show the potential for four (4) cut ends, with two of the four based on consensus among the examiners, and the other two somewhat more speculative.

The examiners have attempted to reconcile the fuel rod fragment lengths that have been located with the 18-inch-long target segment lengths, using the following logic. Some of the end cross sections of the "short" fuel/clad fragments, such as FF027, are potential candidates; this segment is a little over six inches in length and appears to have been at least partially cut on one end cross section, while the opposite end cross section appears to have been separated entirely by ductile tearing in bending or pure shear failure. The failure surfaces are consistent with the damage that might be caused by attempts to force a fuel rod segment into a container by repeated "jamming."

Such jamming loads would not only cause trauma to the end cross sections, but also cause bending loads at or near the separated ends, depending on the angle of the jamming action.

Some of the fuel rod fragments have two very nearly planar/perpendicular cross sections at the separated ends, one of which shows signs of cutting and the other of which shows essentially a shearing failure or a combination of ductile tearing and shear. FF027 is an example of flush fuel pellet with relatively reflective end surfaces for fuel and cladding on one end, with a combination of shear and ductile tearing failure on the other end. FF030 is a 4-l2-inch-long segment that has evidence of cutting on one end and shearing failure on the other end. FF033 and FF034 are two other segments that are three to four inches long, with some evidence of cutting on one end and ductile tearing on the other end.

Some of the fuel rod fragments have two very nearly planar/perpendicular cross sections at the separated ends, both of which appear to be sheared off. FF031 is typical example of this behavior.

10

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 The pertinent fuel fragment descriptions are as follows:

  • FF009. A 30-1/4-inch-long fragment with a top endplug. The fractured surface end does not satisfy any of the five criteria.
  • FF013. An approximately 4-inch long segment with fractures at both ends. There is a circumferential crack that is approximately 1 inch from one end of the fragment. A partial cut mark is located approximately 1/4-inch from the circumferential crack. This segment is a strong candidate for a remnant piece of rod A-49.
  • FF021. A 14-inch-long bottom end piece, with a one-inch split and folded back cross section. The folded back end appears to be planar (see Figure 7). The fuel pellet is not flush, raising the possibility that this might be a circumferential crack instead of a cut.

However, this cannot be ascertained with confidence because the fuel and clad surfaces normal to the potential cut/cracked cross section cannot be viewed.

  • FF026. An 11-inch-long top end remnant. The potential cut cross section satisfies all of the five criteria. Figure 8 is a still shot of the most telling view. The cross section is planar and nearly perpendicular. The fuel is flush with the cladding, and both the fuel and cladding surfaces normal to the cross section are very reflective. This represents the standard against which all other potential cross sections are measured.
  • FF027. A 6-'A-inch-long segment with one end that has been potentially cut and the other end showing a combination of ductile tearing (from lateral bending) and shear tearing failure. The fuel is flush with the cladding on the potentially cut end (see Figure
9) and the exposed fuel surface is very reflective. By Criterion 5, both of these features indicate a cut; however, the exposed cladding surface normal to the cross section is moderately reflective, but dinged up. This is a reasonably strong candidate for one cut end
  • FF028. A one-inch-long fragment that has one reasonably planar end (except for a ductile tearing flap) and one jagged end. The fuel pellets are not flush on either end and are not reflective. This does appear to be a strong candidate for an IGSCC crack, not a cut end.
  • FF029. A one-inch-long fragment with one end reasonably planar and the other end jagged. The fuel pellet on the candidate cut end is nearly flush, but has been subject to end damage that has caused some fuel loss at the end. The cladding has been severely damaged also, to the point that its reflectivity is almost nil. A portion of the exposed fuel surface is very reflective, indicating a cut section. However, it is not a strong case for a cut.
  • FF030. A 4-Y2-inch-long fragment with two planar ends, one of which has fuel flush with the cladding, and one of which is torn in shear. The fuel is very reflective (see Figure 10),

and the cladding is partially reflective, but badly dinged. This is a strong candidate for a cut at one end, where the fuel is flush with the cladding and is reflective, and an IGSCC failure at the other end.

11

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant. March 31, 2005

  • FF031. A 2-Y4-inch-long fragment with one end planar, except for a ductile tear that possibly occurred at the near completion of cutting. The fuel is reasonably flush and reasonably reflective indicative of a cut. The cladding has been badly dinged, to the point that reflectivity is almost nil. This is a good, but not a great candidate for a cut end.
  • FF032. A 4-42-inch-long fragment, neither end of which satisfies any of the five criteria.

This appears to be a fragment that has been torn apart from any cut/cracked ends.

  • FF033. A 3-5/8-inch-long fragment, with one end relatively planar (with a ductile tearing flap), and the other end jagged and sheared. The fragment shows significant permanent bending and some partial tearing from lateral loading. The exposed fuel surface is partially flush, with some loss of material, and relatively reflective on the intact portion of the exposed fuel surface. The cladding is too badly damaged at the potentially cut cross section to determine its reflectivity. This is a potential candidate for a cut end, but the case is very weak for either possibility, cut or IGSCC.
  • FF034. A 3-1/4h-inch-long fragment with two planar ends. One of the ends has an exposed fuel surface that is almost flush as shown in Figure I1 (the cladding surface is too badly damaged to determine), while the other end is torn and jagged, as if from shear failure.

This is a candidate, but not great, for a cut at one end and an IGSCC crack at the other (jagged surface).

Therefore, in addition to a reasonable case that FF026, FF021, and FF013 maybe remnants of the cut A-49 rod, two of the other fragments are strong candidates for having cut ends - FF027 and FF030 - while one fragment - FF031 - is a moderate candidate for a cut end. In addition, three other fragments - FF029, FF033, and FF034 - have some characteristics of cut ends, but the arguments are inconclusive.

12

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Conclusions Still photographs, drawings, and digital videos have been extensively reviewed by an expert review team assembled by ATI Consulting. The judgment of this team, after reviewing all available information, has concluded that there is reasonable evidence consistent with the proposition that fragments from the three 18-inch segments along with the remnants cut from the A49 fuel rod may be amongst the fuel fragments in the HBPP spent fuel pool.

A case can be made for the presence of the end remnants of that fuel rod, which represents two of the cut ends that are being sought. In addition, with respect to the other possible cut ends, a reasonable case can be made for two of those cut ends, and a moderately reasonable case can be made for a third cut end. Thus, the evidence supports that at least one of the 18-inch segments (which have been broken into fragments) may have been found, and by inference the other two segments may also be contained among the rod fragments found in the spent fuel pool. Potential candidates for the remaining cut ends have been identified, but the cases are much weaker. Of the five criteria that were used to identify the potential cut ends, only one or two of the criteria are met; in some instances, the criteria are only partly met because of damage to the potential cut ends, especially to the exposed cladding surfaces normal to the cross section. In spite of this damage, the fact that two of the candidates met all five criteria, and that a third candidate met all of the criteria, at least in part, lend confidence to the findings and conclusions.

13

193Cm

.-P-K ,

TtN¢f; I

2

  • 1. I J

.1 1

I AI C. CS, L I..

[ 1 ! 1 j932G AEt r

2C6053 Fr I -'.

3---

ni IFUEL C

"@tttl ROD .

-n 7. I1AVwATtA$,mtvIItr4Tmptsu AU.

c:.

.A.

M..AALLCmOWAf 2.AmIs

%OW U

0,*e/Oj 4 AJ

. . II TOtCCX 01 Mt Pw.

[M OTht '.A-b F"4 peFLP

11 IIIs I

a 14'AS117 PI CD FAMt"ATMM Kt41JLW SAM.L tJOT *tHXL jO*-t .- L EL, 1.0 PPVI CY Alnomr: _ _l v D P tdSA567Z PI C-

  • e xcrrbzrf wtte r^s n *e, usoqL

'3PorI+7L SHALL86 ___ t5 -iVI8 1 1485334 P:

4 .A~uLAtif 1OF 5M pun AXIJITH ItS-7 I MAHtumw tjFAsL wcF. e A4 7 c tSPUrao TtOet AMIL MIAT WOr 95 s5' OF TH _____.7 CAp 1A 7? PI memb 5' ' PA:.< tr PCLE T

.L3T AtSASS74-ASAS8S4-~I Pi

____a 0

4t tCM INCULS ALcoWAWi " tDb 9. UUMW' I1r's r _ __ !ui PtL~t- tSA-tTP,_ =_ _

2 :e4~4L 4 FILL Au. VoMMWiTHREAC1 A0CPCftIM&

~ TtV$wsr 71f -D FCfr.i97fg%7

6. I1m*tt nlt bWAX." P3LEe pq WH,,

CD M4AY ISC UOSCOa tCCVSPCJP T.

m Tn. .

. tTn il C _

0n f I

  • h  ;
  • L .i II I :D

.. : e ' 5,, uZc ..

.I 0

. -1 ft  :,'.-t. I Fa-Ni I

ii:-

. / I rF=-ts;  ; .Y: "

0Ir Z M I -34!-24 1 -.! :t  ::=lf.@:@:s , .

C 3,

.i1::I. L __. - , . _ . .. _

I-' `0 0.-

r-CD .~ 'T_ . tP. tt w

. ;CA;r rp7 C) ,'

,.urNp' tiLTts

  • I ZIWTc1) m 'n SI)

CD ILi Co CDa. - - -. 4.; >5 )@@St£ 5-?$-*~r 1e1m - _

B n

,*I P-

< ] A/if

. IV _ _

I .- M f'l

- I_ I/ eOBT LI?, tnm.rUnv4I CD CD CD _____ -- 'a--- I I 0

CD (D

CD 0 Co CD CD 03 E Ii Fe,-'-

I~~

I@-.- ~w i --- 1"

,i~Se .-. =. I 4

~

=lS/ S=- °lc- }e{4w~;E*l_C5_.

Hlw 2 t s; ew w.1 rn Ir _ S

.1- -lI l

§ S I .1 - -

'Mn I l 2 1 3 1 4T 1 5 . I

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 t'*3- a 1 Ci -.. j

____ - - = Ia077Oc ry 'MrW

, 1,--' ,- - -.L' , . . .. .- 1 ., - -.. . .i -.-. 41.....

Figure 2. Schematic showing original cutting plan for the corner rod in Assembly A-49.

Shown are the locations of the file marks to be used for segment identification and the order of the cutting procedure from rod bottom to the rod top.

of IGCC Tma.

i~I 1 T~ Noted i iul xm 1 R: in T!i I~~l j

i 11I . IY..VY

.V. , 1 _ - i. ,

.; .1 .1r-

. . i.

I _i_

--- I......

. .. I

'..-I

. -A

_ _ __ _ _ _ _ 4 5: .. - . Z. .. - I ..

I

. s  ;-w=1>r8l7C_8-,SE

______ ~ ~ :Ii.

I X.iwo#.4-.,

j 1

~ 1
  • . r-ask,. .. .S I

I

, -1I

!_:.  ! 2.i.4.I*A:.I.*J IA; j "  !

.  ; f - jI

, I J.."Ic.-i - .... . ;  : . 4 i I Figure 3. Schematic showing alternative cutting plan for the corner rod in Assembly A-

49. Shown are the locations of the file marks used for segment identification and the order of the cutting procedure from rod bottom to the rod top. Approximate locations of the fuel assembly grid spacers and fuel rod damage from visual examinations also are shown. Possible post-cutting condition of the fuel rod from Assembly A-49 is also shown. Postulated locations of fuel fragments FF021, FF026, and FF013 are indicated.

15

ATI consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31. 2005 a

Figure 4. Picture of Fuel Fragment FF026 showing endplug with dimensions consistent of top endplug (shank length is approximately 1.0 inch).

Figure 5. Picture of Fuel Fragment FF021 showing endplug with dimensions consistent of bottom endplug (shank length is approximately 0.5 inch).

16

ATI Consulting Report. Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Possible Partial Cut Figure 6. Pictures of Fuel Fragment FF013 showing evidence of a partial cut.

17

ATI Consulting Report. Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Figure 7. FF021 photo showing folded back cross section.

1" I

".-i El 1- 1.

Figure 8. FF026 photo showing evidence of cut end meeting all criteria.

18

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 d

Figure 9. FF027 photo showing flush end indicative of a cut surface.

. . .. . I

. t-Figure 10. FF030 photo showing reflective surfaces indicative of cutting.

19

6.

ATI Consulting Report, Evaluation of Nuclear Fuel Rod Fragments and Inference to Fuel Rod A-49 at Humboldt Bay Power Plant, March 31, 2005 Figure 11. FF034 photo showing possible cut surface but badly damaged cladding surface.

20

Enclosure 2 PG&E Letter HBL-05-010 RESUMES OF THE INDIVIDUALS WHO PREPARED THE ATI REPORT

Resume WILLIAM L. SERVER Professional Qualifications Mr. Server has thirty years of experience in materials applications and structural integrity evaluations focused primarily in the electric utility industry. At ATI Consulting, Mr. Server is one of the founding principals and is continuing the development of business opportunities and alliances focused on component life cycle management and life assurance procedures for nuclear utilities and other industries. Prior to forming ATI Consulting, Mr. Server specialized in fracture mechanics applications, failure analysis, mechanical/materials testing using state-of-the-art techniques, and technical/management decision-making.

Experience 1991 - President,ATI Consulting. Mr. Server is internationally known for his expertise in the area of reactor pressure boundary aging degradation. Current activities at ATI include life attainment and license renewal strategies, application of advanced fracture mechanics approaches to complex technical issues, reactor pressure vessel (RPV) radiation embrittlement, mitigative actions to materials aging, and decision-making approaches for utility management. Over the last several years, Mr. Server has consulted extensively with: Wisconsin Public Service Corporation on direct measurement of fracture toughness for assessing RPV integrity; First Energy Nuclear Operations Company on application of the Master Curve for assuring extended life operation of the Beaver Valley Unit 1 RPV; EPRI on fundamental understanding of radiation embrittlement mechanisms and fracture mechanics applications; Duke Energy and Niagara Mohawk on development of comprehensive and integrated reactor vessel documentation programs; and Bechtel Bettis on fundamental microstructural characterization of ferritic steels. Other key utilities that Mr. Server has consulted with over the years include Pacific Gas and Electric, Consumers Energy, Baltimore Gas and Electric, Southern California Edison, and Carolina Power and Light. He has continued to assist the nuclear power industry and EPRI in developing embrittlement management programs and completing an integrated international research RPV irradiation/surveillance program. Mr. Server also has led several International Atomic Energy Agency (IAEA) coordinated research programs and expert missions (Argentina and Croatia) representing U.S. commercial interests in radiation damage effects and RPV integrity.

1988 - 1991 SeniorProjectManager, TENERA, L.P. Emphasis was on nuclear plant life extension, fracture mechanics evaluations, radiation embrittlement management and mitigation, and improved reactor vessel surveillance programs. He went on an expert mission for the IAEA to Institut "Jozef Stefon , Lubijana, Yugoslavia, to teach a course on safety aspects of aging phenomena in nuclear power plants.

ATI Consulting I 9116/04

William L. Server Page 2 of 3 1986- 1988 Managerof MaterialsEngineering, Robert L. Cloud & Associates, Inc.

Specialized in fracture applications and analyses, leak-before-break applications, material degradation mechanisms, and fatigue crack growth models and analyses.

1981 - 1986 Managerof Metals PropertiesGroup/SeniorEngineeringSpecialist, Idaho National Engineering Laboratory, EG&G Idaho, Inc. Conducted and supervised fracture mechanics evaluations, annealing studies for reactor pressure vessels, life extension of nuclear components, and evaluation studies of NRC licensing issues.

He taught an undergraduate Materials Engineering and Fracture course at the University of Idaho, Idaho Falls Extension.

1975- 1981 President, Fracture Control Corporation. Managed and conducted static and dynamic mechanical properties of materials and structural components, conducted failure analyses, and developed small specimen fracture mechanics testing/analysis techniques. He went on an expert mission to Comision Nacional de Energia Atomica, Buenos Aires, Argentina, for the IAEA. He also taught introductory Materials Science in the Chemical and Nuclear Engineering Department at the University of California, Santa Barbara.

1972- 1975 Director, MaterialsScience Programns/ManagerofDynatup Services, Effects Technology, Inc. Managed, conducted, and supervised static and dynamic mechanical properties evaluation and testing, acoustic emission applications, fracture mechanics methods, and new products development.

Education B.S. Metallurgical Engineering, Colorado School of Mines (1970)

M.S. Materials Science and Engineering, University of California, Los Angeles (1972)

Partial completion of MBA program, University of Idaho Honors B.S. Degree with High Honors Tau Beta Pi, Sigma Gamma Epsilon, and Blue Key Engineering Honoraries Colorado School of Mines Board of Trustees and International Nickel Company Scholarships UCLA Materials Science Department Fellowship ATI Consulting 2 9116/04

William L. Server Page 3 of 3 Professional Affiliations American Society of MechanicalEngineers (ASME)

Member: ASME Boiler and Pressure Vessel Code Section XI, Past Secretary:

Subgroup on Evaluation Standards; Member: Working Groups on Flaw Evaluation and Operating Plant Criteria; Past Chairman and Member of several special Task Groups.

American Societyfor Materials (ASM)

Past Chairman: Eagle Rock Chapter, Idaho Falls, Idaho American Societyfor Testing and Materials(ASTM)

Member: Committee on Nuclear Technology and Applications (El0),

Subcommittee on Behavior and Use of Nuclear Structural Materials (E10.02);

Chairman: Task Groups on Inservice Annealing (E 509) and Conducting Surveillance Tests (E 185)

Member: Committee on Fatigue and Fracture (E08), Subcommittee on Elastic-Plastic Fracture Mechanics Technology (E08.08); Past Secretary: Task Group on Precracked Charpy Test.

American Nuclear Society (ANS)

Member.

Pressure Vessel Research Council (PVRC)/MaterialsProperty Council (MPQ Member: Task Group on the Toughness Master Curve; Past Member: Working Groups on Reference Toughness and Precracked Charpy Test.

InternationalAtomic Energy Agency (IAEA)

Lead Chief Scientific Investigator for Coordinated Research Program on Surveillance Programs Results Application to RPV Integrity Assessment Publications Published over 100 technical papers, made more than 70 conference presentations, and written over 100 technical and contract reports.

ATI Consulting 3 9/16X04

- 9 ROBERT E. NICKELL Dr. Robert E. Nickell provides engineering consulting services to private industry and government through Applied Science & Technology, a California C corporation based in Poway, California.

His current consulting activities include:

  • EPRI - Technical support to EPRI Project Managers on nuclear power plant license renewal technical issues, in particular reactor water environmental effects on metal component fatigue life and management of aging effects for PWR internals components.
  • EPRI - Technical peer review on nuclear power plant security issues, such as aircraft impact on plant structures housing nuclear fuel and explosive loading on plant structures; high-level support for the development of the U. S.

Department of Energy Nuclear Plant Security Roadmap.

  • Los Alamos National Laboratory - Chair, Confinement and Safety Vessel Design Review Panel for DynEx Project; external review of codes and standards, such as the ASME Code, and the corresponding materials issues and design analysis, for confinement vessels used in the U. S. nuclear stockpile stewardship program of the U. S. Department of Energy.

Other recently completed consulting activities include (with completion date):

  • General Electric Company - External peer reviewer, through the law firm of Morgan, Lewis & Bockius LLP, on safety assessments for Tokyo Electric Power (TEPCO) BWR internals cracking issues (October 2002).
  • Sandia National Laboratories - Structural peer review panel for the nuclear spent fuel transportation Package Performance Study (May 2002).
  • Portland General Electric - Technical support to the PGE staff on Trojan Nuclear Power Plant decommissioning, most recently review of safety analyses for the Holtec dry storage cask system (June 2002).
  • Pacific Gas & Electric Company - Review of technical specifications and dynamic structural analyses (e.g., seismic restraint system for the transfer cask) for the independent dry nuclear spent fuel storage system at the Diablo Canyon Nuclear Power Plant (May 2002).
  • Institute for Regulatory Science - Vice Chair of external peer review panel on mitigation of groundwater contamination at the Nevada Test Site of the U.

S. Department of Energy (December 2001).

  • Kobe Steel, Ltd. - Evaluation of pitting corrosion and stress corrosion cracking in Formosa Petrochemical refinery vessels (March 2002).
  • Gesellsehaft fur Nuklear-Behalter mbH - Design and licensing technical assistance on the CASTOR X/32 spent fuel storage and transport cask Safety

D -

Analysis Reports, and the Kori KN-12 spent fuei transport cask, including material fracture toughness, welding, and inspection issues (June 2001).

Dr. Nickell received his B.S. (1963), M.S. (1964), and Ph.D. (1967) degrees in Engineering Science from the University of California, Berkeley. His early professional career was spent in the Bell System:

  • First, at the Bell Telephone Laboratories, Whippany, New Jersey (1968-1971);
  • Then, on an industrial sabbatical teaching assignment at Brown University, Providence, Rhode Island, as an Associate Professor of Engineering (1971-1973); and.
  • Finally, at the Sandia National Laboratories, Albuquerque, New Mexico (1973-1977).

During this ten-year period, Dr. Nickell conducted and supervised both fundamental and applied research for private industry and government sponsors, serving as Supervisor of Solid Mechanics at the Bell Telephone Laboratories and Supervisor of Design Technology at Sandia National Laboratories.

Since 1977, he has been a private engineering consultant, except for a four-year period (1980-1984) when he was a Project and Program Manager at the Electric Power Research Institute (EPRI), Palo Alto, California, and a three-year period (1992-1995) when he was Technical Director at SGI International, La Jolla, California.

While at Bell Telephone Laboratories, Dr. Nickell directed research and development activity on manufacturing problems of the Western Electric Company, and his staff provided specialty consulting services to Western Electric on structural design and analysis of both commercial and defense systems.

While at Brown University, he taught courses on soil mechanics, structural design, advanced structural dynamics, and finite element methods, while conducting research on residual stresses and deformations in welded structures, viscous fluid flow with free surfaces, nonlinear structural dynamics and buckling, forced convective/conductive heat transfer, and acoustic-structure interaction.

While at Sandia National Laboratories, he directed the management of projects on nuclear spent fuel transportation, ASME Code rules and related standards for elevated temperature reactor design, residual stresses in welded aerospace structures, inservice inspection of nuclear pressure vessels and components, seismic loading simulation using explosives and centrifugal accelerations, structural integrity of PWR component supports, and scale-model light-water-reactor severe accident experiments.

While at EPRI, Dr. Nickell managed research projects on repair welding of heavy-section steel vessels and components, residual stresses in BWR piping, fracture toughness of steam generator and reactor coolant pump support materials, aging of cast austenitic stainless steel components, simplified piping design, new design criteria for flexible piping systems, degradation and failure of bolted joints, and pressurized thermal shock of irradiated PWR reactor vessel weldments. As Program Manager, he directed the budget and technical management of research projects in the areas of structural integrity of nuclear power plant components, design and licensing of advanced light-water-cooled reactors, cleanup activities at Three Mile Island Unit 2, and high-temperature gas-cooled

reactor (HTGR) research. The latter program involved the study of failure modes of dissimilar metal weldments and recrystallization of high-nickel alloy steam generator tubing materials. In addition, he provided support to the EPRI External Fuel Cycle Program in the areas of ductile cast iron and other alternative materials for spent fuel dry storage casks.

Dr. Nickell has been involved in various ASME Boiler and Pressure Vessel Code activities for the past thirty years. He is currently a member of Subgroup NUPACK, ASME Code Subcommittee III, which has been developing rules for the construction of containment systems for nuclear spent fuel and high-level waste transport packagings, and is the past Chair of the Design Working Group of that body. He is currently the Chair, Task Group on Impulsively-Loaded Vessels, reporting to the Special Working Group (SWG) on High Pressure Vessels of Section VIII of the ASME Code. He was the Secretary of the Task Group of Reactor Pressure Vessels as Shipping Containers of Subcommittee XI. He was the elected Chairman of the Consultants Service Meetings (CSMs) that developed criteria for the evaluation of brittle fracture for radioactive material transport packagings, under the auspices of the International Atomic Energy Agency (IAEA). Over the years he has taught short courses on the design of pressure vessels to ASME Code Section VIII, Division 1 and 2 rules, as well as short courses on repairs and alterations to nuclear power plant components.

Dr. Nickell is a member of ASCE, ANS, and ASTM, and is a Fellow of the AAAS and ASME. He is a past Technical Editor of the ASME Transactions Journal of Pressure Vessel Technology, a past Chair of the Executive Committee of the ASME Pressure Vessel & Piping Division, a past member of the Accreditation Board on Engineering &

Technology (ABET) mechanical engineering curriculum accreditation visitors lists (six visits in six years from 1978-1984), a past member of the ASME National Nominating Committee, a past Chair of the ASME Transactions Board of Editors, a past Chair of the ASME Council on Engineering Budget Committee, a past Vice President and Chair of the Board on Communications of the ASME, and a past member of the Finance Committee of the Board of Governors of the ASME. He was elected and served as Governor of the ASME from 1992-1994, and served as Chair of its Committee on Program Review from 1993-1994 and as Chair of its Task Force on International Direction. He served as the Chair of the Board's Committee on Finance and Investment from 1994-1999, monitoring the ASME's annual budget of some $ 70MM and its investment portfolio of some $

80MM. Dr. Nickell was nominated in June 1998 and elected in November 1998 to become the 11 8th President of ASME. He assumed that office in June 1999 and served until June 2000. He is currently a Past President of ASME, and serves as the Secretary and Treasurer of the Society, with a three-year term ending in June 2004.

Dr. Nickell was the 1972 recipient of the Office of Naval Research/American Institute of Aeronautics and Astronautics (ONR/AIAA) Naval Structural Mechanics Award, and was appointed by U.S. Secretary of Energy Hazel Rollins O'Leary to the National Coal Council for 1993-1995, with a reappointment for the periods 1995-1997 and 1997-1999.

He was selected to present the Robert D. Wylie Memorial Lecture at the Ninth International Conference on Pressure Vessel Technology in April 2000. He has authored or co-authored more than 80 papers in refereed journals.

-'
^b RESUME ABSTRA CT Y. R. (Joe) RASHID Dr. Rashid received B. S., M. S., and Ph. D. degrees in Civil Engineering from the University of California, Berkeley, in 1960, 1962, and 1965, respectively.

PROFESSIONAL HISTORY Dr. Rashid founded ANATECH in 1978. He remains its Chairman, Chief Executive Officer, and Technical Leader. Prior to joining ANATECH, from 1970 to 1978, he held the position of Manager, Fuel and Structural Mechanics at General Electric -

Nuclear Energy Division in San Jose. From 1964 to 1970, Dr. Rashid was Manager, Analytical Research and Development at General Atomic in San Diego. During his years as a graduate student at Berkeley, Dr. Rashid served as a Research Assistant for four years.

PROFESSIONAL REGISTRATIONS, SOCIETIES AND HONORS As a result of his extensive background and experience, Dr. Rashid has been involved in advisory work to a number of national and international research institutions and industrial organizations. He is a member of the American Society of Mechanical Engineers (ASME), American Concrete Institute (ACI) and a founding member of the international SMiRT Conference. He holds a Professional Nuclear Engineer License in the State of California. Dr. Rashid has received numerous honors through his services on technical advisory boards and expert panels. Dr. Rashid was elected a Fellow of the American Society of the Mechanical Engineers in 1999.

BACKGROUND Dr. Rashid has over 35 years of experience in the analysis break failure mode for concrete containments, which was of complex structures. As a contributor to the general later verified experimentally by Sandia's reinforced technical literature in engineering and structural concrete model test. He actively participated in the expert mechanics, his work in computational methods are elicitation process of NUREG-1 150. As part of the internationally recognized and span a wide range of overall development effort in reactor containment activities including original development in three- technology programs, he and his team were responsible dimensional finite element modeling and solution for the development of concrete containment failure techniques, irradiated materials characterization, criteria, both deterministic and probabilistic. Similarly computational fracture mechanics constitutive modeling for steel containments and reactor pressure vessels, he in concrete and metal viscoplasticity, analytical modeling developed a strain-based creep rupture failure criterion for of reactor fuel element behavior, large scale computations the TMI vessel, which extends the classical Larson-Miller in nuclear structures, failure analysis of reactor Parameter criterion into the plastic instability range.

containment structures, analysis of spent-fuel facilities, and the development of source-term-based methodology Dr. Rashid has been a Project Manager on scores of for spent fuel shipping casks. He has authored more than projects during his 30-year career. He also has extensive 100 papers and reports in these areas. He was the experience in fluid/structure interaction, particularly as principal developer of several major computer programs. applied to General Electric (GE) Mark I, II & III wet well These include the development of industry-standard blowdown loadings. While a senior manager at GE in the computer codes for nuclear fuel elements (FREY), late 1970's, Dr. Rashid led a team of in-house staff and concrete structures (ANACAP), and advanced-reactor contractors that studied and resolved many of the structures (SAFE series of codes). important licensing issues related to the blowdown steam bubble collapse in the GE Mark I BWR suppression pool.

Dr. Rashid is a structural/mechanical/nuclear engineer and a leading contributor to the development of reactor Dr. Rashid has participated in or had an opportunity to plant safety evaluation methodologies. His expertise in follow along with virtually all major developments in the behavior of containinents and pressure vessels under computational mechanics over the last three decades. He severe accidents is well established in the nuclear participated in many "firsts": the development of the first industry. He served as chairman of the Structural finite element program for the analysis of three-Mechanics Peer Review Group for the TMI Vessel dimensional solids; one of the first computational Investigation Project conducted by NRC and OECD. plasticity models; the development of the only nuclear Based on his pioneering research with prestressed reactor fuel safety (predictive) analysis program (FREY);

concrete vessel technology in the late 1960's, he was the the development of the first robust plain, reinforced, and first to promote the now-accepted concept of leak-before- prestressed concrete predictive analysis computer ANATECH mmu Iii

RE UE- BR C RESUAIE ABSTRA CT Y. R. (Joe) RASHID programs (ANACAP and ANAMAT); and the development of a rebar bond slip model including the Bauschinger plasticity effect and cyclic degradation. In addition, Dr. Rashid has been involved in many unique applications of this methodology to challenging problems of national and international interest. These include: the validation of the ANACAP/ANAMAT concrete constitutive model on the NRC/Sandia 1:6 scale model test of a reinforced concrete containment structure and on the NRC/Sandia/British 1:10 scale model test of a prestressed concrete containment (failure pressure, failure mode, failure location and various loading histories were accurately predicted pretest); and the blind pretest predictions of the failure loading, failure mode and location, and various loading histories for two Caltrans highway bridge "bent-cap" scale models.

Dr. Rashid is the leading expert in cask drop evaluation and is the author of the well-known EPRI Target Hardness Methodology which demonstrates that cask deceleration loads for realistic target are bounded, regardless of drop heights, by the energy absorption capacity of the target. The monographs and models of the target hardness methodology and the analysis tools used to develop these monographs have been used extensively by many utilities for the licensing of spent fuel casks and storage facilities.

ANATECH

_ _ _ _== !!

r- rr

RESUDfEABSTRAC ROBERTO. MONTGOMERY EDUCATION:

Mr. Robert Montgomery received B. S. degree in Nuclear Engineering from Texas A&M University, 1984 and received his M. S. degree in Nuclear Engineering from Texas A&M University, 1987 PROFESSIONAL HISTORY Mr. Montgomery joined ANATECH in 1988 as Associate Research Engineer and in 1993 was appointed to Manager Nuclear Technologies. Prior to joining ANATECH, from Sept. 1986 to Nov. 1986, Mr. Montgomery served as a Research Assistance with BELGONUCLEARIRE, Brussels, Belgium.

BACKGROUND Tables (PIRT) review document for the BWR Anticipated Mr. Montgomery, currently Vice President and Manager Transient Without Scram (ATWS) Power Oscillation of ANATECH's Nuclear Technology Group, has 20 years Event. He also supported the PIRT review for the PWR experience in the modeling and analysis of nuclear fuel Control Rod Ejection Event. Mr. Montgomery has been rod behavior under normal operation, transient conditions, involved with the assessment of irradiation effects on and in a defective state. He has performed root cause transient fuel behavior during a Loss of Coolant Accident evaluations of fuel reliability issues including cladding (LOCA). This assessment has included analytical and failure by PCI, manufacturing defects, and cladding technical support of experimental programs at Argonne corrosion. In support of utility fuel performance National Laboratory, the Halden Reactor Project, and the assessments, Mr. Montgomery has evaluated the effects Japanese Atomic Energy Research Institute (JAERI). Mr.

of irradiation and operating conditions on cladding failure Montgomery has extensive experience in the high during power changes and transient events. Under his temperature behavior of zirconium 'alloys and uranium leadership, the Nuclear Technology Group at ANATECH dioxide pellets applicable to LOCA conditions. Mr.

has performed evaluations of power operating restrictions Montgomery has been actively supporting the Nuclear and developed recommendations for improving Industry's effort to develop an industry-wide approach to operational flexibility of nuclear plants without licensing high burnup fuel. He has performed a review compromising the reliability of nuclear fuel performance. and assessment of burnup effects on the fuel rod acceptance criteria used in the design and licensing of Mr. Montgomery was a key member of the nuclear fuel.

EPRI/GE/Industry Task Force on BWR fuel secondary degradation in the early 1990's. He was the project leader Mr. Montgomery has extensive experience in modeling of in the development of analytical methods for the nuclear fuel rod thermal, mechanical, and chemical modeling and analysis of BWR fuel secondary behavior under normal operating and accident conditions.

degradation. These activities included integrating He has developed material and constitutive models for use experimental results from laboratory tests, reviewing and in fuel rod behavior codes. These models include, assessing off-gas activity levels from operating material oxidation and hydriding, mechanical experience, and interpreting post-irradiation examination performance, fission gas release, heat transfer, and results. Based on his experience in BWR secondary thermal-hydraulics. Mr. Montgomery has a familiarity degradation, Mr. Montgomery developed BWR fuel with a variety of fuel rod analysis codes, including FREY, operating guidelines to mitigate the consequences of ESCORE, DEFECT, FALCON, SCANAIR, and secondary degradation. Mr. Montgomery has also served TRANSURANUS. In addition, he has performed on the EPRI/NEI Task Force on Reactivity Initiated thermal-hydraulics calculations using RETRAN.

Accidents (RIA's) and has been the primary technical consultant to this task force on the analysis of RIA Mr. Montgomery has written more than 30 technical experiments performed throughout the world. Mr. publications related to nuclear fuel rod behavior.

Montgomery was the project manager for the technical development of revised regulatory acceptance criteria used in the licensing of control rod ejection accidents (CEA) in PWRs. He was the primary author of the topical report recently submitted to the NRC for a generic review and approval for the PWR CEA event. Mr.

Montgomery was a member of the NRC Experts Panel that developed the Phenomena Identification and Ranking

A- o-RU
1 RESUMIEABSTRA CT DIONJ. SUNDERLAND EDUCATION Mr. Sunderland received a B.S. degree in Nuclear Engineering from Texas A&M University in 1982. After completing his thesis research on Mixed (UPu) Carbide Sphere-Pac Fuel, he received an M.S. degree in Nuclear Engineering from Texas A&M University in 1985.

PROFESSIONAL HISTORY Mr. Sunderland has broad-based experience in Light Water Reactor (LWR) nuclear fuel and core component technologies. In May 1998, Mr. Sunderland joined ANATECH and is supporting programs in nuclear fuel performance modeling and nuclear power safety sponsored by the Electric Power Research Institute and the Nuclear Regulatory Commission. Prior to working with ANATECH, Mr. Sunderland was a SeniorNuclear Engineer with StollerNuclear Fuel, a Division of NAC International.

At Stoller, Mr. Sunderland specialized in fabrication quality assurance, design analysis, performance assessment, and failure and root-cause analysis of LWR nuclear fuel and core components. During graduate school, Mr. Sunderland's work focused on nuclear fuels modeling, high temperature material behavior, and space nuclear power and propulsion systems applications for interplanetary spacecraft.

PROFESSIONAL REGISTRATIONS, SOCIETIES AND HONORS Mr. Sunderland is currently a member of the American Society for Metals, ASM International (1985-present), the Institute of Electrical and Electronics Engineers (1986 - present), and the American Society for Testing and Materials (1998). During his years at university, Mr. Sunderland was a student member of the American Institute of Aeronautics and Astronautics (1986-1988) and American Nuclear Society (1979 - 1988)

BACKGROUND stresses. Other activities include use of COMETHE and Mr. Sunderland has broad experience in LWR nuclear ESCORE to evaluated burnup effects such as fuel fuel technology with over ten years of projects involving: temperature, fission gas release and rod internal pressure.

  • auditing of the fabrication of LWR nuclear fuel and In the area of fuel mechanical analysis, Mr. Sunderland core components, such as control blades, was a principal investigator in various design reviews of advanced BWR fuel designs for lead assembly programs
  • fuel design analysis for new fuel designs, or the (LUAs or LTAs) and reloads. Such work has involved modification of existing designs, mechanical design reviews of GEl I for Detroit Edison
  • fuel performance assessment, including a and SPC 9x9-9QA for the Washington Public Power comprehensive database of LWR fuel failures in the Supply System. The review of the SPC 9x9 design US and Europe, included the calculation by Stoller of the stresses in tie-rod endplugs under design loads, and a review of an SPC
  • assessment and root-cause analyses of failed nuclear (Exxon) tie rod tensile test report.

fuel, and

  • assessment of the mechanisms and degradation In the area of PWR mechanical analysis, Mr. Sunderland behavior of failed nuclear fuel. supervised the application of finite element modeling of PWR grids in order to assess the mechanical performance These projects were performed on behalf of individual of grid springs in various duel designs including and groups of utilities in the US, Europe and Asia, as well Westinghouse V5H mid-grids and bottom Inconel grids, as EPRI and the US Department of Energy, ORNL. Siemens bimetallic grids and Fragema AGA-2G grids.

He has performed basic design analyses of fuel rod At Stoller, Mr. Sunderland performed several fuel internal pressure, flow induced vibration, etc. for most performance simulation projects, including the analysis of modern PWR fuel designs used in Europe and the US.

BWR fuel ramp tests using the COMETHE III-K fuel modeling code. Using this code, he performed From 1996 - 1998, Mr. Sunderland provided technical simulations on a variety of 8x8, 9x9, and l0xl0 BWR support to ORNL as part of DoE's Fuel Material fuel designs from GE, SPC (ANF, Exxon), and ABB and Disposition Program in which MOX fuel is being analyzed failure thresholds based on Zircaloy cladding developed from Pu-239 taken from nuclear warheads. As part of this work, Mr. Sunderland developed a draft MOX

i' RESUAME AVSTRA CT DIONJ. SUANDERLAND pellet specification and a draft process outline for the investigator in three utility root-cause investigations pellet manufacture. Additional work involved an involving PWR fuel at Wolf Creek (WCNOC), Palisades assessment of fuel swelling and fission gas release of (Consumers Power Co.), and TMI-1 (GPU Nuclear). A MOX test fuel irradiated in ATR experiments. fourth project involved assistance in the root cause analysis of failed toe-rod endplugs in old Exxon 7x7 fuel As a complement to work in fuel performance and design at Oyster Creek (GPU Nuclear).

analysis, Mr. Sunderland has extensive experience auditing LWR fuel. This work has involved numerous Mr. Sunderland has been a principal investigator and technical fabrication and process-inspection audits of both contributor in the evaluation of failures in BWR barrier BWR and PWR fuel assemblies and components, BWR and non-barrier fuel. In a series. of three projects channels and control rods. Mr. Sunderland has sponsored by BWR utilities and EPRI, Mr. Sunderland participated in audits at ABB-CE (Hematite), BNFL conducted case studies of individual fuel failures. These (Springfields, UK), GE (Wilmington), SPC (Richland), case studies involved the collection and interpretation of ANFI (Lingen, Germany), and Westinghouse (Columbia). off-gas and coolant activity data in order to identify the initial time of failure and on-set of degradation.

Fuel designs audited include: Analytical work involved evaluations of the fuel rod power history and stress calculations in cladding for

  • ABB's SVEA-96 (for WPPSS), several cases in order to assess the potential for failure by
  • BNFL AGR fuel (for Nuclear Electric, UK), pellet-cladding interaction (PCI). In support of the mechanical analysis, Mr. Sunderland compiled and
  • GE10, GE12 and GE14 (for EGLKKL, TVO and reviewed literature on the oxidation of U0 2 and RWE, and BKW/KKM respectively) oxidation, hydriding and fracture of Zircaloy.
  • SPC 8x8 and 9x9 BWR fuel (for RWE/BAG/

Gundremmingen), The collection and review of literature regarding the mechanical properties of Zircaloy and Zircaloy fracture

  • SPC 16x16 PWR fuel (for RWE/Biblis), are on-going as part of Mr. Sunderland's work at Stoller.
  • Westinghouse 17x17 OFA (for Taiwan Power/ As a complement to this work, Mr. Sunderland completed Maanshan) a detailed study for Tokyo Electric Power Co. involving a evaluation and comparison of the current cladding
  • Westinghouse VVER-I 000 (for CEZ/ Temelin) materials offered by ABBA (LK2+ and LK3), GE (Process 6) and Siemens (LTP-2). The ABBA portion of Particular audits have focussed on the fabrication and this project included a review of the evolution of QA/QC of: ABBA/Sandvik cladding from LKO through LK3.
  • Fuel pellets (UO 2 and U0 2 -Gd 2 O3 )

Based on his extensive knowledge of LWR nuclear fuel

  • Zircaloy-2 and -4 components, fabrication, design, performance and failure analysis, Mr.

Sunderland has provided training seminars to utility

  • tubing (TEX) and tubing for cladding, water rods, personnel on these subjects. In 1998, Mr. Sunderland was and guide tubes invited to provide training on the fabrication of PWR fuel
  • bar stock and endplugs, to a team of engineers and junior managers at the Kansai Electric Power Co., Osaka, Japan. Other seminars have
  • sheet and strip, spacers and channels, been provided to TVA (1995), Washington Public Power Supply System (1992), and a utility group (Consumers
  • BWR Control Blades Power Company, Philadelphia Electric Company,
  • AGR Pellet, Fuel Rods and Assemblies American Electric Power Company, Indiana and Michigan Power Company, Niagara Mohawk) in 1990.

The bulk of Mr. Sunderland's recent activities have been in the areas of fuel failure analysis. He was a principal

Retrieved from "https://kanterella.com/w/index.php?title=HBL-05-010,_ATI_Consulting_Report_Evaluation_of_Nuclear_Fuel_Rod_Fragments_and_Inference_to_Fuel_Rod_A-49_at_Humboldt_Bay_Power_Plant&oldid=2593770"